Investigation of Feasibility of Incorporation of Hydride in Fuels

Advanced Test Reactor National Scientific User Facility (ATR NSUF)

Donald Olander1, Kurt Terrani1,4, Tom Newton2, Gordon Kohse2, Lin-wen Hu2,

David Carpenter2

Mitch Meyer3 , Jim Cole3 , Joy Rempe3

1 University of California, Berkeley2 Massachusetts Institute of Technology

3Idaho National Laboratory4Oak Ridge National Laboratory

Work supported by the U. S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-051D14517, as part of an ATR National Scientific User Facility experiment.

Outline

o Introduction to LWR Hydride Fuel with liquid metal gap filled Concept

o Laboratory Experiment

o Irradiation Experiment

• Temperature, power measurement

• Thermal conductivity deduction

• Cover gas analysis

o Proposed post Irradiation experiments

o Summary

2

Liquid-metal-bonded fuel rod concept

pellet stack

Cladding

He LMSpring

End plug

He

gap

UO2 U-ZrH1.6

Conventional fuel rod Proposed fuel rod

4

What is a hydride fuel?TRIGA fuel: uranium metal + zirconium hydride - (U,Zr)H1.6

up to 45 wt % or 21 vol% a-U dispersed in -ZrH1.6 matrix

10 m

Limiting U content

Why so much U?

U density of hydride only 40% that of UO2

For the same linear power, need ~ 10% enriched in U235

NRC regs?

Black: U

Gray: ZrH1.6

(U,Th,Zr)Hx probably a better fuel

Hydride fuel+ Liquid-metal Gap: A Better LWR Fuel?

But: can these two technologies be combined in fuel rods for LWRs?

• Space-Nuclear-Auxiliary Power (SNAP) Program NASA 1960 –70• TRIGA Research reactors – since 1957• Control rod for fast reactors: U.S. Navy

Metal Hydrides

Liquid Metals• Sodium-cooled fast reactor EBR II, Phenix, JOYO• Lead-cooled fast reactor (Gen IV)

Why Hydride in Place of Oxide?• Moderator (H) is in the fuel => reduce volume of

water => smaller core, pressure vessel – or, higher power from same core

• Higher LHR possible – higher burnup (enrichment limited)

• Improved safety: faster negative feedback than oxide fuel (TRIGA)

• Lower fuel temp. khyd~ 6 x kox (Tmax < 650oC) - reduced FP release - reduced stored energy

Thermal conditions in LWR rodshydride hydride oxide

characteristic He bond LM bond He bondpellet OD, mm 10 10 10

LHR, W/cm 375 375 375

fuel centerline, oC 680 555 1505

Tfuel 170 170 995

Tgap (35 m) 125 1 125

Tclad 46 46 46

Tfluid 39 39 39

coolant, oC 300 300 300

Fission-gas release from U-ZrH1.6

Note: data are old (1960 – 1980) and poorly documented

9

Swelling of U-ZrH1.6

SNAP program (1965) data

What causes this?

Fission- product swelling 3x that of UO2 !!!

Fuel-cladding chemical interaction• Zircaloy is a powerful sink for hydrogen available

from the fuel

At 700oC: pH2(fuel) ~ 10-1 atm; pH2(cladding) ~10-4 atmfuelcladding

Fuel cracking?• Thermal stress – tensile at periphery• Hydrogen redistribution – transports H from the center to the surface; generates compression at the surface

Hydrogen redistribution in temperature gradient

Total stress is compressive at periphery – prevents cracking?

11

SEM versus AFM Elastic modulus mapping across the microstructure of

(U4Th2Zr9)H1.5 fuel

100

120

140

160

180

200

220

0 0.5 1 1.5 2 2.5 3 3.5

Position (m)

240

200

150

100 100

120

140

160

180

200

220

0 2 4 6 8

Ela

stic

Mod

ulus

(G

Pa)

Position (μm) Position (μm)

•-U phase modulus is 210 GPa•-ZrH 1.6 phase modulus is 125 GPa

•The elastic modulus of ThZr2H7-x phase is determined 172GPa for the first time12

Modified AFMBackscattering

SEM

Compatibility with Zircaloy cladding• Zircaloy samples pressed against hydride

fuel at various contact pressures immersed in liquid metal at 375 ºC for one month

1 mm

20 μm

Zircaloy

LMHydride

Zircaloy

LM

Hydride

1 mm

Zircaloy

LM

Hydride

Stainless Steel

200 μm

Zircaloy

Hydride

20 µ

m g

ap

120

MPa

con

tact

TERRANI, K., et al., “Liquid Metal as a Gap Filler to Protect Zircaloy Cladding from Hydride Fuel,” Proceedings of Top Fuel 2009, Paris, France

τ

F

Torque Screw

Stainless SteelZircaloy

Pt Wires Hydride Fuel

-

Irradiation Experiment

Objectives• Design, construct and irradiate a mini-fuel

element under realistic LWR conditions– At stack midplane:

• maintain Tmax ~600oC

• >10% U-235 burnup• gap closure

– At ends of stack• Tmax ~ 500oC

• gap remains open– On-line temperature read outs– on-line fission-gas monitoring

~12 cm

fuel

LM

clad

Pellet FabricationTR

IGA

Fuel

Sl

ugDi

amon

d Co

re D

rills

Roug

h Pe

llets

Po

st D

rillin

g

U(30wt%)-ZrH1.6

19.7% U-235

Grinding Wheel

Rubber Wheel

Adjustable Arm

Pellet Position

Stepper Motor

Smoo

th P

elle

ts

Post

Grin

ding

Cent

erle

ss G

rindi

ng

Appa

ratu

s

Mini Fuel element assembly

Thermal conductivity measured by:

•Two thermocouples; fuel centerline and rod surface

Sheath TC Welded to SS Flange

SS304 CF Mini Flange

Zr CF Mini Flange

He PlenumSS302 Spring

Pb-Bi Alloy

Alumina Spacer

Zircaloy-2 Tube

U0.17ZrH1.6 Fuel

Zircaloy-2 End Cap1 cm

Neutron radiography at MITR with resolution of ~ 100 μm

Active Fuel RegionAlumina Spacers

Zirconium FlangeSS 304 Flange

302 SS Spring

Rod 1

Rod 2

Rod 3

Rod 4

Ti

Cover gas line

  

   

 

 

 

 

 

         

     

  

  

 

      

 

 Center TC

Pellet

LM

Capsule 3 (Rod 2

Capsule 2 (Rod 3)FCT-3 and CST-3)

Capsule 1 (Rod 1)FCT-1 and CST-1

Flux profile

Core StructureCore Tank

Typical Fuel Element

Fine-Control Regulating Rod

Shim Blade Absorbers (6)

Shim Blade Flow Relief Holes (6)

Hexagon Strut(No Absorbers)

Coolant Entrance Channels (6)

Fixed Absorber in Radial Strut

A-2

A-1A-3

B-1

B-2

B-3

B-4B-5

B-6

B-7

B-8 B-9C-1

C-2

C-3

C-4

C-5

C-6C-7C-8

C-10

C-11

C-12

C-13 C-14 C-15

10 cm

MIT Reactor Core

• Assemblies simultaneously irradiated at each time • One assembly removed every 4 months (Burnup dependent data) • Longest assembly to remain within the core for 1 year (0.30%

Fission of initial metal atom, FIMA)

Test matrix for the HYFI irradiation of fuel rods

Irradiation Position

Irradiation dates

March 20, 2011

May 26, 2011

June 25, 2011

Nov. 11. 2011

Jan 20, 2012

Early 2012

Top Dummy Dummy Fuel rod 2 Fuel rod 2 Fuel rod 2 Fuel rod 4(w /thermal cond. Probe)

Middle Fuel rod 3 Fuel rod 3 Dummy Dummy Dummy Fuel rod 5(W He filled gap)

Bottom Fuel rod 1 Fuel rod 1 Fuel rod 1 Fuel rod 1 Dummy Fuel rod 2

Abandoned

Temperature and thermal power profile of fuel rods 1, 2 and 3 since March 2011

Note: Unpredicted frequent shot downs and ramp ups

Thermal conductivity calculation through annual fuel pellet

Neutronic (MCNP) calculations

For 5 MW:

300

310

320

330

340

350

Power LHR

Fuel Portion

Pow

er [W

/cm

3 ]

200

210

220

230

240

LHR

[W/cm

]

H,

Rr0

T0

Ts TTC

TLMcool

Ti

Oxidized call

Time dependence of the thermal conductivity

0.155

0.16

0.165

0.17

0.175

0.18

k12k34

Date

k did not change with at the beginning of irradiation

Lower capsule thermal conductivity estimate

0

200

400

600

800

1000

1200

0.16 0.165 0.17 0.175 0.18

k (lower cap), W-cm-1T-1

21ΔT

(W)th

P5.70

Lower k

k12

Minimum 0.15856169Maximum 0.1770716Points 14244Mean 0.1701022Median 0.17003919Std 0.0011548277d

Thermal conductivity variation of three fuel rods during irradiation

0.1

0.12

0.14

0.16

0.18

0.2

0 500 1000 1500 2000 2500 3000 3500

Capsule 1 (rod 1)Capsule 2 (rod 3)Capsule 3 (rod 2)

Time (hr)

• The initial rise is attributed to the lag time of thermal power with respect to TC readouts

• What is the reduction of deduced thermal conductivity due ?:• Large initial swelling• Good retention of fission gas products• Hydrogen redistribution in the fuel• Or Oxidation of clad, formation of bubbles in LM, configurations changes with

time within the duct holding capsules

Thermal conductivity change on Capsule 3 by Jan.

2012

The drastic reduction in deduced-conductivity is of concern. Needs to be verified by post irradiation analysis

Fission products release due to possible leak in capsule 1

Burnup estimate

Neutronic (MCNP) calculations for 5MW:

Upper Solid

Upper Annular

Middle Solid

Middle Annular

Lower Solid

Lower Annular

300

310

320

330

340

350

Power LHR

Fuel Portion

Pow

er [W

/cm

3 ]

200

210

220

230

240

LH

R [W

/cm]H,

Comparison of In-pile Thermal Conductivity of U(30wt%)-ZrH1.6 and

UO2

Structural Damage at MIT NRL

On January 24, 2012Damage to bottom spacer and is identified, some missing alignment pins on capsules are All parts except bottom spacer moved to storage locations

On November 2011Increasing fission gas release from Capsule 1 triggers its removal, is replaced with the dummy from wet storage

Capsule 1

Bottom Spacer

It has been decided to terminate the irradiation at this time and send the three capsules to INL for post irradiation analysis

PIE to determine irradiation effects on fuel & cladding

• Fission-gas behavior• Bubbles in LM bond - state in fuel

• Fission-product swelling• Verify SNAP data

• Fuel-cladding chemical interaction (hydriding)

• Can LM protect Zircaloy cladding from attack by fuel?

• Fuel cracking• Do pellet chips in gap stress cladding? - Do cracked wedges close

gap?

31

Post-irradiation examination

Phenomenon Instrumentation

Fuel cracking OM, SEM

Xe in the plenum gas Mass Spec.

Gas bubbles in the frozen alloy OM

Hydrogen distribution in fuel SIMS

Hydride precipitates in cladding OM, SEM

Uranium particles SEM, AFM

Void around U particles TEM, AFM

Diametral expansion of fuel and cladding

Micrometer

Second-phase particles in fuel SEM, AFM

Summary• Irradiation of hydride fuels has started in March 2011

and terminated January 2012

• Clad and centerline temperature, reactor power as well as fission products release were monitored in time

• Out of 5 mini rods prepared, three were actually irradiated

• Post irradiation analysis will be accomplished at INL

References•K. Terrani, J. Seifried, D. Olander, “Transient Hydride Fuel Behavior in LWRs,” J. Nuc. Mat., 392, (2009) 192.•D.R. Olander, E. Greenspan, H.D. Garkisch, B. Petrovic, “Uranium-zirconium hydride fuel properties,” Nucl. Eng. Design, 239, (2009) 1406.•Kurt A. Terrani, Mehdi Balooch, Gordon Kohse, David Carpenter, Lin-wen Hu, Mitchell K. Meyer, Donald Olander “In-Pile Thermal Conductivity Measurement of Uranium-Zirconium Hydride Fuel” Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors (NFSM 2012) June 24–28, 2012 , Chicago, IL